1. Field of the Invention
The present invention relates to a process for treating radioactive waste liquid and, more particularly, to a treatment of concentrated radioactive waste liquid containing a soluble salt as the main component.
2. Description of the Prior Art
A solid of liquid low-level radioactive waste formed in a nuclear power plant or other facilities using radioactive rays is packed in a drum, where it is solidified with cement and is stored in this state in the nuclear power plant because delivery of the radioactive waste outside the nuclear power plant is not allowed. The volume reduction ratio of the radioactive waste is low in this cement solidification method and the number of cement-solidified drums stored in facilities increases year by year. Accordingly, from the viewpoint of economy of the storage space, various methods for increasing the volume reduction ratio of the radioactive waste have been developed.
As the main liquid radioactive waste formed in the nuclear power plant, there can be mentioned a concentrated radioactive waste liquid. At the present, this concentrated radioactive waste liquid is dried and powdered to remove water occupying the majority of the volume, and solidification is effected with a hydraulic solidifying material, especially cement. However, the following problems arise when the concentrated radioactive waste liquid is dried, powdered and then solidified with the hydraulic solidifying material.
In case of a concentrated radioactive waste liquid generated in a BWR power plant (wherein the main component is sodium sulfate (Na.sub.2 SO.sub.4)), calcium hydroxide (Ca(OH).sub.2) deposited when cement is set reacts with sodium sulfate (Na.sub.2 SO.sub.4) to form calcium sulfate (CaSO.sub.4), which in turn reacts with tricalcium aluminate (3CaO.Al.sub.2 O.sub.3) hydrate to form a swellable hydrate which deteriorates the solidified body.
In case of a concentrated radioactive waste liquid generated in a PWR power plant (wherein the main component is sodium borate (Na.sub.2 B.sub.4 O.sub.7)), if the amount of the borate ion is increased relative to cement, the hydration of cement is hindered and a desirable solidified body cannot be obtained. This problem arises not only when cement is used but also when other hydraulic solidifying material is used.
As means for solving the foregoing problems, there has been examined a method in which a concentrated radioactive waste liquid is subjected to an insolubilizing treatment (wherein sodium sulfate or sodium borate (Na.sub.2 SO.sub.4 or Na.sub.2 B.sub.4 O.sub.7), the soluble component in the concentrated radioactive waste liquid, is converted into an insoluble substance), and then, the concentrated radioactive waste liquid is dried, powdered and solidified. Particularly, insolubilization of the concentrated radioactive waste liquid generated in the PWR power plant is disclosed in Japanese Patent Laid-Open No. 186099/1983.
However, from the results of experiments made by inventors of the present invention, it was found that powdering and solidification of the concentrated radioactive waste liquid are difficult if caustic soda (NaOH) formed in the insolubilizing treatment of the concentrated radioactive waste liquid is contained in the concentrated radioactive waste liquid to be subjected to powdering and solidification.